The types of fuel loaded and burned in domestic nuclear power plants are WH-type and OPR/ APR-type nuclear power plants, with a total of 19 types. In the case of spent nuclear fuel released in Korea, the low combustion level of 45,000 MWD/MTU or less accounts for about 75%. In terms of fuel type, WH 17×17 and CE 16×16 fuels account for about 85% of all spent nuclear fuels. The thickness of the oxide film of the fuel cladding can make the fuel rod vulnerable during reactor operation, directly affecting the integrity of the fuel rods. so, it is a very important design factor in design. Therefore, the fuel rod design code that predicts and evaluates this has also been developed to accurately predict fuel rod corrosion. And it’s being applied to the design. In this study, the ECT probe measured by inserting it between fuel rods. The thickness of the fuel cladding oxide film was measured for spent nuclear fuel. When reloading operational nuclear fuel, the IAEA recommends an oxide film thickness of up to 100 micrometers. In this study, it was confirmed that spent nuclear fuels keeping integrity burned for 2-3 cycles were sufficiently maintained within the limit. However, the difference could be confirmed according to the characteristics of the coating material, the combustion cycle, and the use of poison rods. For the reliability of the data, symmetrical to the quadrant fuels were selected, and the fuel burned at the same period was measured. The method of selecting the target fuel can produce meaningful results.

Spent nuclear fuels are temporarily stored in nuclear power plant site. When a problem such as cracking of spent nuclear fuel assembly or cladding occurs or uranium that has not been separated during the reprocessing remains, it is necessary to treat it. The borosilicate glasses have been considered to vitrify whole spent nuclear fuel assembly. However, a large amount of Pb addition was necessary to oxidize metals in assembly to make them suitable for oxide glass vitrifcation. Furthermore, these borosilicate glasses need to be melted at high temperatures (> 1,400°C) when UO2 content is more than 20wt%. Iron phosphate glasses can be melted at a relatively low temperature (< 1,300°C) even with a similar UO2 addition. A composition of iron phosphate glass for immobilization of uranium oxide has been developed. The glasses have glass transition temperatures of ~555°C that are high enough to maintain its phase stability in geological repositories. The waste loading of UO2 in the glass is ~33.73wt%. Normalized elemental releases from the product consistency test were well below the US regulation of 2 g/m2. Nuclear criticality safety and heat generation in deep geological repositories were calculated using MCNP and computational fluid dynamics simulation, respectively. The glass had effective neutron multiplication factor (keff) of 0.755, which is smaller than the nuclear- criticality safety regulation of 0.95. Surface temperature of the disposal canister is expected to lower than the limit temperature (< 100°C). Most of the U in the glass is in the 4+state, which is more chemically durable than the 6+state. As a result of long-term dissolution experiment, chemically-durable uranium pyrophosphate (UP2O7) crystals were formed.

Thermodynamically, TRUOx, REOx, and SrOx can be chlorinated using ammonium chloride (NH4Cl) as a chlorinating agent, whereas uranium oxides (U3O8 and UO2) remain in the oxide form. In the preliminary experiments of this study, U3O8 and CeO2 are reacted separately with NH4Cl at 623 K in a sealed reactor. CeO2 is highly reactive with NH4Cl and becomes chlorinated into CeCl3. The chlorination yield ranges from 96% to 100%. By contrast, U3O8 remains as UO2 even after chlorination. We produced U/REOx- and U/SrOx-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH4Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REOx and SrOx can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrOx-simulated fuels.

상용원자로에서 발생하는 산화물 사용후핵연료의 부피감용과 재활용을 위하여 산화물을 금속으로 환원시 키는 공정에 대한 연구가 수행되어 왔다. 다양한 환원법 중에서, 한국원자력연구원은 LiCl-Li2O 용융염을 반 응매질로 사용하는 전해환원공정을 현재 개발 중이다. 파이로 공정의 전단부에 해당하는 전해환원 공정은 PWR 산화물 연료 주기를 소듐냉각 고속로의 금속연료 주기에 연결시켜 준다. 이 논문은 금속전환 공정을 개 발/개선하고, 용량 증대를 수행한 한국원자력연구원의 노력을 요약한다.

경막결정화를 이용한 산화물 사용후연료의 전해환원 공정에서 발생하는 LiCl 염폐기물 내 포함되어 있는 Cs 및 Sr을 분리(농축)에 대한 실험을 수행하였다. 결정화 공정에서 Cs 및 Sr과 같은 불순물들은 불순물들의 용융염 상 및 결정상에 대한 용해도이 차리로 분리되어 최종적으로 작은 양의 LiCl 용융염내에 농축된다. 본 연구에서는 LiCl-CsCl-SrCl2 계에대한 고체-액체 상평형도를 통해 결정화를 통한 분리가능성을 파악하였으며 열전달방정식 의 계산을 통해 경막결정화 운전중 LiCl 용융염상의 온도분포를 예측할 수 있었다. 경막결정화 공정에서 결정성 장 속도는 분리효율에 큰 영향을 미쳤으며 90%의 LiCl 재생율을 가정할 경우 20-25 l/min의 냉각속도 그리고 0.2 g/min·cm2보다 작은 결정성장 속도조건에서 각각의 Cs 및 Sr에 대하여 90% 정도의 분리효율을 나타내었다.